This invention relates to a method of supervizing the degree of thermo-hydrodynamic stability of the coolant flow in the fuel channels of the reactor core of a nuclear reactor of the type in which boiling of the coolant is permitted in the reactor core, for example a boiling water type nuclear reactor, and more particularly to a method of supervizing the stability of the coolant flow in the fuel assemblies by determining the coolant flow quantity in each fuel assembly which is an important factor for determining a stability limit by signals produced by neutron flux detectors installed in the reactor core.
The coolant passing through the fuel channels each consisting of a fuel rod bundle enclosed by a channel box in the reactor core of a boiling water type nuclear reactor removes the heat generated in the fuel bundle while flowing through the fuel assemblies in the form of a two-phase flow of water and steam. As is well known in the art there is a tendency of hydrodynamic oscillation of the two-phase flow caused by feedback between voides, pressure loss and flow quantities i.e. density wave oscillation. For examle, R. T. Lahey, Jr. and F. J. Moody explain the classification of dynamic instabilities of interest in boiling water type nuclear reactors as density wave oscillations, pressure drop oscillations and flow regime induced instability in their book, "The Thermal Hydraustatic of a Boiling Water Nuclear Reactor", published by the American Nuclear Society, 1977. However, from the point of view of the reactor designer, the most important dynamic instability that must be considered is the density wave oscillation, so that hereinafter, we will consider it. Since the reactor core is provided with a plurality of fuel assemblies which form parallel flow passages, hydrodynamic oscillation is created in each fuel assembly independently of other fuel assemblies under a condition wherein the pressure difference between its inlet and outlet ports is constant. Furthermore, in view of local nature of the oscillation, it can be considered that the feedback through reactor kinetics is not significant so that such feedback can be ignored in the first approximation. Such oscillation mode is termed "channel stability" whereas an oscillation mode wherein the reactor kinetics and the dynamic characteristic of the recirculation flow path are related to the feedback phenomena of the two-phase flow in the fuel assemblies is termed "reactor stability".
The reactor stability during the actual running of a nuclear electric power generating station can, from its definition, be supervised by watching signals regarding the neutron flux in the core, recirculation flow quantity, etc., whereas since the channel stability concerns a local phenomenon it is not always possible to supervise the same by observing only the output of the neutron flux detectors in the core. For this reason, it has been the practice to determine a range of stable operation by carefully analyzing the channel stability at the time of designing the reactor.
The operating conditions of a nuclear electric power generating station may vary during running over an extended period depending upon the type of the fuel, and various limits on the operating conditions. In such a case, the stability is analyzed for modifying the operation range. There are also small changes which do not require such modification. For this reason, it is correct to consider that the limit of the channel stability of a nuclear reactor in actual operation varies from time to time. Since the operating range at the design stage is determined by considering most serious conditions there is considerable allowance of the stability in most cases. Accordingly, the freedom of operation could be greatly increased if the limit of the channel stability under actual operating states were determined by actually measured data in all fuel assemblies.